The present invention relates to a process for the separation of large amounts of uranium from small amounts of radioactive fission products, which are present in basic, aqueous carbonate containing solutions, by means of an organic, basic anion exchanger.
Until now, in order to recycle irradiated nuclear fuel elements from compounds, or alloys of highly enriched uranium, respectively, nuclear reactor fuel elements were dissolved in nitric acid and the uranium separated by liquid/liquid extraction, as for example in the Purex process, or by amine extraction, or by column chromatography separation operations, and reprocessed in a nitric acid medium.
The nitric acid recycling of nuclear fuels, especially the Purex process, is a reliable process that has been known for a long time. Nevertheless, it is extremely problematic that targets cooled for a short time (for example, cooling periods of 1 to 30 days) can be reprocessed with nitric acid. The disadvantages of nitric acid reprocessing of targets which have cooled for a short time are as follows:
The presence of the shorter lived fission products, especially iodine-131 and xenon-133, make the use of holdback systems or delay beds, respectively, extremely necessary. With the use of nitric acid (other acids cannot be used because of their corrosivity) and the associated possibility of developing NO.sub.2, the most effective and also most economical filter material, activated carbon, may not be applied, because otherwise, in case NO.sub.2 is released, there would be an acute danger of combustion in the waste gas lines.
Further, all fluid/fluid extraction processes are especially difficult to manage for high grade systems charged with I-131 and Xe-133 (as in this case), because, along with the danger of Xe-133 emissions, there is the additional possibility, which has considerably more serious consequences, of HI and iodine emissions from the acidic system.
A further disadvantage of the fluid/fluid extraction is the increased expenditure necessary to avoid the danger of combustion caused by the extraction agent diluent. The use of noncombustible diluents, such as carbon tetrachloride, is not recommended in this extremely highly active system because of the pronounced radiation sensitivity and the increased danger of corrosion by the released hydrochloric acid.
In addition, all efficient extraction chromatography processes known until now occur in acid systems and have, along with the previously cited disadvantage of the HI and I.sub.2 release, respectively, an additional great disadvantage, that is the fixing of uranium from the main portion in the process stream, with reduced holdback of the fission products. The disadvantage of this method is quite obvious: For nuclear fuel holdback, incomparably larger column volumes must be prepared.
It is known to reprocess uranium dioxide, or alkali diuranate residues of high U-235-enrichment, respectively, extremely contaminated with fission products such as one obtained after the alkaline decomposition of material-test-reactor-fuel elements. The elements consist preponderantly of a uranium/aluminum alloy of the approximate composition UAl.sub.3, coated with aluminum. Because of the variable Al content in the compound, the designation UAl.sub.x is usually used. This fuel element type is frequently established as the starting target for the production of fission product nuclides for nuclear medicine and technology. For that, usually smaller elements are exposed by thermal neutron streams of about 1.times.10.sup.14 n/sec cm.sup.2 for 5 to 10 days. In order to minimize loss of the desired nuclide by decay, the irradiated targets are transported to the reprocessing installation after a minimum cooling time of about 12 hours. Usually, an alkaline decomposition of the target with 3 to 6 molar soda lye, or potash lye, respectively, serves as the first chemical step. In this first chemical step, the main constituent of the plate, the aluminum, and the fission products soluble in this medium, such as the alkaline and alkaline earth ions, as well as antimony, iodine, tellurium, tin and molybdenum, go into the solution, while the volatile fission products, above all xenon, together with hydrogen formed from the Al solution, leave the solvent at the upper end of the reflux cooler. Hydrogen can be oxidized to water over CuO, while xenon is preferably held back at normal temperature on activated carbon delay beds. The non-spent uranium remains as insoluble residue, usually about 99% of the initially irradiated amount, as UO.sub.2 or alkali diuranate, respectively, together with the insoluble fission product species, above all ruthenium, zirconium, niobium and lanthanides in the form of their oxides.
This residue is treated in a known method with the action of air or of an oxidation agent, as, for example, H.sub.2 O.sub.2 or hypochlorite, with an aqueous, carbonate- and hydrogen carbonate-ion containing solution of pH 5 to pH 11. The concentration of the carbonate ions can reach a maximum of 2.5 m/l and that of the hydrogen carbonate ions a maximum of about 1.0 m/l. During this treatment, the oxides of the uranium and of the named fission product species enter the solution as carbonato-complexes.
For purposes of economy and safety, this briefly cooled, extremely contaminated nuclear fuel must be recycled, retargeted and then stored. The usual method with nitric acid solution, however, is excluded for reprocessing briefly cooled fuel elements on a technically achievable scale, as already explained, because of the raised iodine-131 contamination even after the treatment, as well as the known combustion danger of the activated carbon in the presence of nitrogen oxides.